Neutronic reactor



July 25, 1961 R. F. CHRlSTY 2,993,852

NEUTRONIC REACTOR Original Filed Oct. 19, 1945 10 Sheets-Sheet 1 Fig.

INVENTOR.

Robert /-T Chrisiy Wl: NESSES:

July 25, 1961 R. F. CHRISTY 7 2,993,852

NEUTRONIC REACTOR Original Filed Oct. 19, 1945 10 Sheets-Sheet 2 W/TA/ESSES: INVENT Robe/f E Chr/ BY H WW July 25, 1961 R. F. CHRISTY2,993,852

NEUTRONIC REACTOR Original Filed Oct. 19, 1945 10 Sheets-Sheet 5INVENTOR. Robert C/m'sfy July 25, 1961 R. F. CHRISTY 2,993,852

NEUTRONIC REACTOR Original Filed Oct. 19, 1945 10 Sheets-Sheet 5 ReactorRoam Control Room A if Release Control 0 o 46 In Out I i 7 /44 A Rad RMagnet Safety Roo' y lay /4/ 239:) 145 rl' ea I09 Density T K /56 IE5Temp nc-oc Source 7g0 Pan Level 0 i----- fl Manometer Level P "MP-E HighT Solution W/ 7'A/E55E5.- INVENTQR.

' Robert E Chr/sty July 25, 1961 R. F. CHRISTY 2,993,852

NEUTRONIC REACTOR Original Filed Oct. 19, 1945 10 Sheets-Sheet 6 I76 UI76 ll ll /72 w/ A/ESSES. mv mozz.

,W M Robert I? Chrisfy W %M44W July 25, 1961 R. F. CHRISTY 2,993,852

NEUTRONIC REACTOR Original Filed 001:. 19, 1945 10 Sheets-Sheet 7 V W IF2s 26 IN VENTOR.

Robert E Chrisfy BY H K' fl/M QWM July 25, 1961 R. F. CHRISTY 2,993,852

NEUTRQNIC REACTOR Original Filed Oct. 19, '1945 10 Sheets-Sheet 8 Fig..2

GRAMS OF 235 0 4 8 l2 I6 20 24 2a l Fly. /2 NCHES U E /o 3 i o m 4 2 InN) N J Ll. O I n g f w l 0 0.01 0.02 0.03 0.04 RECIPROCAL PERIOD(SECONDS W/ TNESSES. INVENTOR. MW w Haber) E Chrisfy r BY 1 W w jgi July25, 1961 R. F. CHRISTY 2,993,852

NEUTRONIC REACTOR Original Filed Oct. 19, 1945 10 Sheets-Sheet 9INVENTOR. W/ T/VESSES f 4 Robert F. Chr/sfy W M W BY July 25, 1961 R. F.CHRISTY 2,993,852

NEUTRONIC REACTOR Original Filed Oct. 19. 1945 10 Sheets-Sheet 10W/T/VESSES: INVENTOR.

Robert E Chr/sfy WM BY H a" M fl%"' 2,993,852 NEUTRONIC REACTOR RobertF. Christy, Pasadena, Calif., assignor to the United States of Americaas represented by the United tates Atomic Energy Commission Originalapplication Oct. '19, 1945, Ser. No. 623,363,

now Patent No. 2,843,543, dated July 15, 1958. Divrded and thisapplication Jan. 13, 1958, Ser. No.

2 Claims. 01. 204-1932 subject matter was divided from said parentapplication pursuant to a requirement for an election of species, noclaim to the generic invention embracing both said unreflected reactorspecies and the reflected reactor species of said parent case beingallowed.

It is generally known that certain atomic nuclei will undergo fissionupon absorption of a slow neutron and will yield through this processtwo nuclei, the sum of whose atomic numbers is approximately equal tothe atomic number of the original nucleus. However, a mass defect existsand a consideration of the process in terms of the conservation ofenergy reveals that a substantial amount of energy is released duringthe fission process. Furthermore, on the average, more than one fastneutron is emitted tor every neutron absorbed to initiate fission.

It is therefore clear that, if the fast neutrons produced by fission canbe made to cause new fissions in such proportion that the overallneutron generation exceeds the overall losses in and from the system,the chain reaction can be divergent to a desired rate of neutrongeneration. As a consequence, the energy released during the fissionprocess is available in the form of heat and/or radiation for extendedperiods of time; that is, during the continuance of the chain reaction.The employment of that energy for useful purposes forms the basis ofthis invention.

The secondary neutrons produced by the fissioning of a fissionableisotope nucleus have a high average energy. More specifically the meanenergy in the fission neutron spectrum is in the neighborhood of from0.5 to 3 million electron volts (mev.), and the mean free path of such'neutrons in a substantially solid mass of fissionable isotope iscomparatively short, for example, of the order of five centimeters, theresult being that the mean time between fissions in such an arrangementwill be of the order of a hundredth of a microsecond. While a fastneutron'chain reaction can be maintained in such an arrangement if asufiicient quantity of such a fissionable isotope or material is broughttogether in favorable geometry, i.e., a quantity in excess of thecritical mass value, it has been determined that for the purposes of thepresent invention, the employment of thermal neutrons to pro ducefissions permits of a number of advantages.

.The requirement that the neutrons employed in a controlled neutronchain'reaction of the type contemplated by this invention be slowed tonear thermal energies by passing them through a slowing medium called amoderator in which they are slowed by atomic collisions, arises out ofthe following considerations: 7

Aswill be shown inmore detail later, the cross section for fission (i.e.the probability that fission will occur under- Patented July 25, 1961neutron bombardment) increases as the energy of the primary or incidentneutron is reduced. In fact, the cross section is approximatelyinversely proportional to the neutron energy. As a concomitant factor,the various moderators have diflierent efficiencies as neutron slowingmedia as well as different absorption cross sections for neutrons. Bythe proper choice of a moderating material, that is one in which theneutrons are quickly slowed to thermal energies and which absorb veryfew neutrons, it is possible to take advantage of the increased crosssection for fission of the fissionable isotope and thereby reduce by asmuch as a factor of ten the critical mass value or the quantity ofmaterial necessary for a selfsustaining chain reaction.

Furthermore, in a fast neutron chain reaction, neutron generation takesplace in extremely short periods of time and neutron density risesexponentially with time, thus presenting control problems which arecomplicated in solution. If on the other hand, the fission neutrons canbe slowed down to thermal energies and a chain reaction initiated, sincethe mean time between fissions in such a reaction will be great,sufficient control over the reaction can be readily maintained and thedesired rate of neutron generation fixed at any desired level.

Isotopes that have been determined to be appropriate for slow neutronchain reaction include, for example, isotopes of uranium (element 92)having the atomic weights 233 and 235 and isotope of plutonium (element94) having the atomic weight 239. These fissionable isotopes have nosubstantial threshold for the energy of the incident neutron, hencefission may be initiated by a slow or thermal neutron, i.e., a neutronwhose energy is aproximately that of thermal agitation.

It might be noted also that, for a substantial part of the energyspectrum, the cross-section 'for fission for these isotopes is almostinversely proportional to the incident or primary neutron energy, thatis, the cross-see tion approximately follows the law. Various mixturesof these isotopes in elemental or compound form and mixtures with otherelements or isotopes can be used when following the teachings of thepresent invention, as will be explained hereinafter.

Fermi and Szilard in U.S. Patent No. 2,708,656, issued May 17, 1955,have disclosed methods and means or establishing slow neutron chainreactions which continue in a self-sustaining manner at predeterminedlevels of neutron density. The system there disclosed provided tor theemployment of uranium in its normal polyisotopic state, that is, uranium238 admixed with approximately 0.7 percent of uranium 235, as thefissionable material. Other component elements which form what is nowknown as a neutronic reactor system include:

(1) A neutron slowing material, known as a moderator, suchas graphite inwhich the fissionable material is dispersed in a geometrical patterndesigned to reduce neutron losses.

(2) Heat removal means for example, channels in heat exchangerelationship with the reacting mass and through which a suitable coolantis circulated in order to stabilize temperatures in the system. 7

(3) An outer casing which serves to reflect neutrons back into thesystem and thereby reduce the quantity (i.e. the critical mass) offissionable mixture necessary to sustain the reaction. This outer casingis sometimes termed a tamper.

(4) Means for charging the reactive elements into the zone in which thereaction takes place and for removal therefrom of the products of thereaction.

(5) A protective shield is sometimes provided around the reactor tominimize the escape of biologically harmful radiations. Such shields maycomprise, for example, bismuth or lead which have been found effectivein stopping gamma radiation, hydrogenous materials such as paratfin forabsorbing neutrons and/or a massive outer concrete casing.

(6) A monitoring system to determine the reaction conditions at alltimes.

(7) Control devices generally comprising neutron absorbing materialsinsertable into the reactive mass to maintain an average state ofneutron production and absorption balance at a predetermined level.

(8) A safety device comprising a quantity of neutron absorbing materialwhich may be used to stop the reaction in case of emergency by beingautomatically inserted into neutron absorbing relationship with thereacting mass.

In considering the requirements for an operating neutronic reactor, theratio of secondary neutrons produced by the fissions to the originalnumber of primary neutrons of the type required to initiate the fissionsin a chain reacting system of infinite size using specific materials iscalled the reproduction or multiplication factor of the system. Thefactor is a dimensionless constant and is denoted by the symbol k. If kis made sufliciently greater than unity to create a net gain in neutronsover all interior losses, and the system is of proper size so that thisgain is not entirely lost by leakage from the exterior surface of thesystem, then a self-sustaining chain reacting system can be built, togenerate neutrons and to produce power in the form of heat by nucleartission.

Important losses of neutrons within the reacting mass have been found tobe by absorption in contaminating impurities which are present with thefissionable mixture (e.g. polyisotopic uranium) or by absorption inuranium 238 without producing fission but instead, leading to theformation of plutonium 239 as will be explained later. The absorption bythe contaminating materials varies, but the effect on the k factor maybe readily determined by the employment of formulae disclosed in theabove mentioned application. The effect of numerous elements has beencorrelated in this way with the composition of fissionable material andmoderator or neutron slowing material. That is, for example, more normalpolyisotopic uranium can be added to a particular system to overcome theabsorption effects of impurities in the system.

Uranium 238 has an especially strong absorbing power for neutrons whichhave been slowed to moderate energies. The energy levels at which thisabsorption is strongest are known as resonance energies, and the neutroncapture or absorption by uranium 238 nuclei at these energies istherefore known as the uranium resonance capture or absorption. Suchabsorption is to be distinguished from absorption in impurities asdiscussed above.

These two neutron loss factors are most important in the determinationof whether a self-sustaining chain reac tion can be maintained. Togetherwith the loss of neutron by leakage out of the system, the abovementioned losses govern the size of the reactor. Thus reactorsconstructed according to prior art principles have been comparativelylarge, massive installations requiring extremely large quantities of thevarious elements and/0r materials described above.

It should also be noted that the efficiency of a neutron slowingmaterial or moderator, depends'upon the scattering cross section of thematerial and its atomic weight. Thus, for example, hydrogen has a highscattering crosssection and a low atomic weight and is an extremelydesirableneutron slowing agent because of the small number of atomiccollisions necessary to slow a neutron to thermal energies. When presentin the form of water however, the absorption cross-section iscomparatively high and the k factors for uranium and water are very 4close to unity and the advantages of the use of the hydrogen are largelylost.

It has been pointed out above that control means are provided inreactors for stabilizing the neutron density at predetermined levels.Such controls have normally been in the form of neutron absorbingmaterials inserted directly into the reacting mass, thus effectivelytaking the neutrons directly out of the reaction. Such controls aresubjected to a great deal of fast neutron as well as thermal neutronbombardment and means for cooling them have been found useful if notcompletely essential.

It will thus be seen that this invention has as an object the provisionof a method and means for establishing a self-sustaining slow neutronchain reaction in a compact unit suitable for general use.

It is a further object of the present invention to provide a means andmethod of so co-relating the essential physical requirements of afission chain reaction that practical and easily controllable neu-tronicreactors can be built.

It is a still further object of the present invention to provide meansfor producing neutrons and radiations for transmutation purposes.

Another object of the present invention is to provide a novel method andmeans for controlling a self-sustaining slow neutron chain reaction.

Another object of the present invention is to provide a reactor systemin which the multiplication factor is independent of most of the neutronlosses generally encountered in such a system.

Other objects and advantages will become apparent from the discussion inthis specification and from the detailed description of illustrativeembodiments which are given by way of example and should not beinterpreted to be limitations of the broad principles underlying theinvention.

The above mentioned objects and advantages are attained by employing acomposition of a fissionable isotope and moderator in fluid form, forexample, one in which the fissionable isotope is suspended or preferablydissolved in a liquid moderator such as water or heavy water (i.e.deuterium oxide, D 0). In such an arrangen'ient the amount of thefissionable isotope present, to a large extent, governs the reaction andeliminates the problems attendant upon complex impurity removaltechniques and the like. In other words, by the use of the methods andprinciples herein disclosed, the neutron absorption effect caused by (a)the presence of impurities, (b) isotopes which absorb neutrons withoutresulting in fission, (0) absorption in the moderator, (d) absorption byfission products and like eifects, can be readily overcome by the novelexpedient of increasing the concentration of the specific fissionableisotope present in the system. Thus, higher neutron losses can betolerated than is the case when natural poly-isotopic uranium is used,but losses still can be overcome to the end that a self-sustaining chainreaction can be maintained. As a consequence, the size of the reactor isno longer a critical factor, the new criterion being the concentrationof the fissionable isotope.

It has been noted that among the materials which can be employed in thepracticelof the present invention are the uranium isotopes of mass 233and 235 and the plutonium isotope 239, all of which have no substantialthreshold for the energy of the incident neutron. These isotopes can beobtained in highlyconcentrated form by isotopic separation procedures orchemical methods (depending on the isotope or element) and brief mentionis made here of such methods as background for this invention and toemphasize further benefits derived from following the novel methods andusing the apparatus herein described.

The fissionable isotope uranium 235 maybe obtained in several Ways.Isotope separation devices such as a mass spectro-separator, similar inoperation to a mass spectrograph butv with larger ion currents, havebeen found satisfactory. Another method of separating the uranium 235isotope from the naturally occurring isotopic mixture is by gasdiffusion methods employing uranium hexafluoride gas and diifusionbarriers. In both methods the separation is not completed in a singlestage, but rather proceeds step-wise, or in cascade fashion, with theaccepted portion of each step being further separated and the rejectedportion being recycled. It will thus be seen that the fissionableisotope is observed to occur in greater abundance or concentration, witheach advancing step in the process.

Uranium 233 may be formed by subjecting a quantity of thorium (element90) 232 to neutron bombardment, the resulting reaction being as follows:

23.5 min. 90233 -)91233+fi 27.4 days 912as 92 aa+ If desired, theuranium 233 can be separated from the thorium parent by chemical methodsbut as will be seen from the discussion herein, this separation is notnecessary if the concentration of the uranium isotope is sufficientlyhigh according to the standards hereinafter set forth.

Plutonium 239 is formed principally by irradiation of uranium 238 withneutrons. As a production method, one way of subjecting large quantitiesof uranium to a high neutron flux is the employment of a reactor such asis disclosed in the above-mentioned patent of Fermi and Szilard. Thereaction leading to the formation of plutonium 239 is:

Since the plutonium is formed in the original uranium slugs dispersed inthe graphite reactor, chemical extraction \and/ or precipitationprocesses may be used to obtain the isotope 239 in a substantially purestate, but here again complete separation is not necessary from thestandpoint of the present invention.

It is generally preferred in the practice of the present invention toemploy a water soluble salt containing the fissionable isotope in thedesired isotopic concentration or in a substantially pure isotopicstate. For example, uranyl salts of high water solubility such as uranylnitrate, uranyl sulphate or uranyl fluoride, plutonium salts such asplutonyl sulphate (PuO SO plutanyl nitrate plutonous nitrate (Pu(NO etc.may be dissolved in water and used in the neutron chain reaction hereincontemplated.

It will be apparent to one skilled in the art that by employing acomposition of a material enriched in a fissionable isotope with a watermoderator, and following the practices and standards hereinafter setforth, it is possible to vary the neutron gain (that is, vary themultiplication factor k) by, increasing the concentration of thefissionable isotope in a given volume. It has been determined that theneutron losses due to the presence of an absorbing isotope such asuranium 238 can be made relatively unimportant without eliminating theuranium 238 fromthe system. Thus, if an isotopic mixture of uranium 235and uranium 238 is employed, if the concentration of uranium 235 issufficiently above that of natural uranium, the losses due to absorptionof neutrons by the uranium 238 become negligible and can be neglected inthe design of a reactor, particularly one using a moderator of highneutron absorption properties such as water even though the amount ofuranium 238'is high. It has been determined further, that where theconcentration of uranium 235 is above about one percent and preferablyabove five percent by Weight of the uranium present, a great reductionin the amount of uranium in optimum geometry necessary to establish achain reaction (i.e. the critical mass of uranium required) can beefiected, and complete or substantially complete removal of absorbingisotopes or impurities is unnecessary.

For example, if a water moderator is used and the fissionable mixture isthe normal isotopic uranium mixture containing 0.7 percent of uranium235, the quantity of material (in the most favorable geometry) necessaryto sustain a chain reaction is extremely largeif a chain reaction can beestablished at all. \By way of comparison, if the enrichment is (theuranium 235 content about two percent of the uranium present), onlyabout 1.7 tons of uranium are required under the same conditions ofoperation. Even more striking is the determination that when the uraniumcomposition contains fifteen percent of uranium 235 a chain reaction canbe maintained, under similar conditions, when only a few kilograms ofthe composition are used. Further reductions of these critical massvalues can be secured through the use of neutron reflectors to cut downleakage losses, but the use of such reflectors does not affect thegeneral principles here noted. An unreflected reactor is referred to andknown in the art as a bare reactor.

The critical mass values for a reactor of substantially sphericalgeometry, as well as the critical dimensions and concentration of thefissionable isotope and the interdependence of these criteria forfissionable isotopes such as have been mentioned, may be calculated asfollows:

The neutron distribution in a reactor as a function of the radius of thereactor is the solution of the diifusion equation:

An+ j n=0 1 where n is the neutron density, An where A is the Laplacianoperator is defined by the relation:

J at an O1: Dy 52 Ari sin kr r then K L =kP (K)'1 where K is a constant.

Let the concentration of fissionable isotope be measured by thermalabsorption by the fissionable isotope per unit volume thermal absorptionby moderator per unit volume Then 1 MM) X) (3) 5 st:( V

where L is the thermal diifusion length of the neutrons in the puremoderator and (1,,(M) and v (M) are respectively the thermal neutronabsorption and thermal neutron scattering cross-sections of themoderator. The

second term in the denominator is a usually negligible correction to thetotal cross-section. It is assumed that the presence of the fissionableisotope does not appreciably change the number of hydrogen nuclei percubic centimeter of the reacting solution. Also where a (M) and a (F)are the fast neutron scattering cross-section of the moderator and theabsorption (and hence the fissioning) cross-section of the fissionableisotopes respectively, a (F) is the absorption cross-section of thefissionable isotope for thermal neutrons and V is the actual number ofneutrons produced per fission. The term fast fission includes the rangewhere the fission cross-section is essentially constant, i.e., from10,000 e.v. up to fission energies. Or stated another way, it wasassumed that the fast fission cross section of about 1 barn (l cm?)remained constant down to an energy E, expressed in electron volts anddefined by E where is the mean natural logarithmic energy decrement percollision in the moderating medium, P (K) is the average probability ofescaping leakage for these energies. Then 8A M EK WM gives theprobability that a collision results in fission,

at( am X is only a measure of the concentration of fissionable isotope.

Substituting (3), (4), (5), in (2) the result is 1 UM Y 5- at( ai( E SAMM K L 5 M Expanding the denominator on the left, one gets a quadraticequation for X.

V(V- 1) 5.5 ,,(111) 03,,(F) E sf( nt( P (K) P,(K) X When Fermis conceptof neutron age applies in the slowing down procedure, so that thedistribution of nascent thermal neutrons from a point source of fastneutrons can be written in which Ms a distance from the source', then P(K) =e and V 1 plosive violence.

'r is the neutron age which is /6 of the mean square displacement of aneutron from place of birth to the point at which the neutron reachesthe energy for which the computations are to be made T1 is theappropriate age of the fast neutrons making fast fission and is therange of the neutron for the first few collisions. In water, thedistribution of energetic neutrons from a fission source is After thefirst few collisions, the distribution spreads in an approximatelyGaussian manner with an age 7' from this lower energy to thermalenergies. This consideration leads to tan Kl tan Kl MK) Kl Kl e TheEquation 7 for X is solved for various values of K. Then the density ofa fissionable isotope such as plutonium 239 which is proportional to Xis known as a function of the critical dimensions of the mixture. For asphere and P (K) for a cylinder of infinite length and for a slab thethickness This permits calculation of the critical mass, mass/cm, andmass/cm. of plutonium 239, for example, respectively for a sphere,cylinder, and slab, as a function of the density of plutonium 239, or asa function of the dimensions.

Except for the region of large density, the critical mass of uranium 235or uranium 233 is greater than that of plutonium 239 by the factor i.e.,by 1.7 or 1 for the same dimensions of the mixture.

Since the function of'the moderating medium, i.e. water, heavy water (D0) or the other low atomic number element having a low capturecross-section, is to slow the fission neutrons, the critical size willbe of the order of the slowing down distance. The minimum concentrationis such that only one of the 2.13 effective neutrons per absorption in1a uranium 235 nucleus and 1.98 effective neutrons per absorption in aplutonium 239 nucleus is absorbed by the chain reactive fissionableisotope compound, the thermal neutron absorption by the fissionablematerial will then be about equal to that by the moderator; the optimumconcentration (minimum critical mass in a sphere) will be about threetimes this minimum. The control of a neutronic reactor is an import-antfactor, since if the reaction is permitted to occur at an unduly rapidrate the reaction will take place with ex- Control of a neutronicreaction may be effected by variation of one or more ofthe above lossesor by variation in the concentration of fissionable ly in the form ofcontrol'rods.

' isotope.

In accordance with the present invention it has been found that aneutronic reactor may be effectively controlled by variation of theleakage from the reacting composition. Thus a neutronic reactor has beenconstructed which is below critical size, i.e., the size of the reactoris so small that leakage of neutrons from the reacting compositionwithout a reflector is too great and this loss alone preventsestablishment of a neutronic self-sustaining reaction. But when thisreactor is provided with a reflector which reflects enough neutrons backinto the reaction zone to reduce the leakage loss, a point can bereached such that a self-sustaining neutron chain reaction can beestablished. The reflector is also provided with means to vary theamount of neutrons so reflected. For example such a means may compriseone or more neutron absorbing control rods which may be removablyinserted in the reflector to absorb neutrons therein. As another meansto accomplish this purpose, for example, a portion of the reflector maybe blocked ofi by neutron absorbers if desired or the amount ofreflector or its depth may be varied. At all events the reactor may becontrolled by control of the leakage factor which may be defined as thedifference between the number of new trons per fission leaking from thereaction zone and the number of neutrons per fission which are returnedto the reaction zone after leaking. This feature of the invention isapplicable to neutronic reactors generally.

In order that the significance of a control by neutron absorbingimpurities be more fully understood, the mechanism of fission will bediscussed further. Not all of the fast neutrons originating in thefission process are emitted immediately. Each chain reacting system hasa characteristic time for neutron generation based upon the percent ofenrichment of fissionable isotope employed in the composition with themoderator, the type of moderator, the reflector used and the like. Thischaracteristic time may be used as a base to which may be related thedetermination of whether the neutrons emitted in the fission process areprompt or delayed. In the fission of uranium 235 about one percent maybe termed delayed neutrons, although the percentage varies for thedifferent isotopes. These delayed fast neutrons may appear at any timeup to several minutes after the fission has occurred. In uranium 235 forexample, half of these neutrons are emitted Within six seconds and 0.9within 45 seconds. The mean time of delayed emission is about seconds.The neutron reproduction cycle is completed by 99 percent of theneutrons in about 0.00003 second in a fluid type reactor systememploying a water moderator such as forms the basis of the presentinvention, although the dependence of this value on the moderator chosenshould be noted. But if the reactor is operating with a reproductionratio near unity, the extra one percent may make all the differencebetween an increase or a decrease in the activity of the reactor. Thefact that the last neutron in the cycle is held back, as it were,imparts a slowness of response by the reactor system to the changes inthe control means that would not be present if the fission neutrons wereall emitted instantaneously.

For cases in which the reproduction ratio (R) differs from unity byappreciably less than one percent, the rise of neutron density, or morespecific-ally the value N to which the number of neutrons has risen froman original value N after a lapse of time of t seconds during and beforewhich the pile has operated at a fixed value of R (N being the number ofneutrons at the beginning of t, i.e., after disappearance of transienteffects due to any preceding change in R), is given by N=N e where Inthis formula a is a fraction of the neutrons that are delayed, erg, inthe case of the uranium 235 isotope a=0.0067. T is the mean delay timefor the delayed neutrons which is in the neighborhood of five seconds inthe case of the same isotope and R is the reproduction 10 ratio of thesystem. The above formula is only approximate and applicable for lowvalues of R because it uses an average delay time.

As an example, suppose as a result of moving the control rod R becomes1.001, and assume that the system has settled down to a steadyexponential rise in neutron density, then that is, N/N =2.72 in 28.5seconds. Hence, doubling of the neutron density occurs about every 20seconds and continues indefinitely. The above formula thus indicates therate of rise for relatively low values of R and shows how the reductionof the rate of the delayed neutron effect is particularly significant inthe stated lower range of R values. Strictly speaking, the givenequation holds only for the steady state, i.e., where R has been heldconstant for some time; an additional transient term must be included toobtain an accurate representation of the neutron density during thefirst few seconds after a sudden change of R.

If R were made exactly 1.0067, a more detailed theory shows that theneutron density would be more than tripled each second. However, if thereproduction ratio R is several percent greater than unity, so that theone percent delayed neutrons are unimportant compared with R-l, thedensity increases at a much more rapid rate as given approximately by(R0.()067)t/l where l is 0.00003 second, the normal time to complete acycle in a reactor such as is described hereinafter. Thus if R were tobe made 1.04, the neutron density would increase in 0.03 second by afactor of approximately 10 over its original level. However, if R were1.02 and 1.03, the factor by which the neutron density would bemultiplied each second, would be 1100 and 700,000 respectively. It isthus apparent that too high a reproduction ratio in a practical systemleads to the necessity of inserting what may be considered as anexcessive amount of controlling absorbers to reduce the efiectivereproduction ratio to unity. An exceedingly dangerous condition couldexist if by accident these absorbers were suddenly completely removed,as the time required for reinserting the absorbing material might be toolong to prevent destruction of the system. As the same eventual densitycan be obtained with a reproduction ratio only slightly over unity, aswith a higher ratio, only at a slower rate, the lower reproductionratios which exceed unity by not substantially more than about 0.01, oran amount equal to the percentage of the neutrons formed which aredelayed neutrons are preferred in practice in the interest of safety.

The application of the principles set forth just above to neutronicreactors utilizing high concentrations of fissionable isotopes, will bemore fully understood by reference to the drawings wherein a preferredembodiment of the present invention is shown in the form of twoneutronic reactors utilizing as the reactive composition therein aqueous(H O) solutions of uranyl sulphate (UO SO containing about 14.6 percentof isotope uranium 235 instead of 0.7% as in natural uranium.

In the drawings:

FIG. 1 is a vertical View partly in section and partly in elevation of aneutronic reactor which has been con structed and is adapted to operateat a one Watt output illustrating the present invention;

FIG. 2 is a cross-sectional view taken as indicated by the line 22 inFIG. 1;

FIG. 3 is an enlarged longitudinal sectional view of the control rodshown in FIG. 1;

'FIGS. 4, 5 and 6 are cross-sectional views, as indicated, of thecontrol rod; 'FIG. 7 is a diagram showing the solution handling system,and also illustrates that embodiment of the present invention in whichno neutron reflector is employed;

FIG. 8 is a diagram showing the electrical control and monitoringsystem;

FIG. 9 is a fragmentary diagrammatic side view partly in section of thetemperature control system and reactor room;

FIG. 10 is a diagrammatic view partly in top plan and partly in sectionof the reactor room;

FIG. 11 is a chart or graph showing the relation between the addedamount of uranium- 235 and the depth of insertion of the control rod;

FIG. 12 is a chart or graph showing the effect of added uranium 235 onthe reciprocal period;

FIG. 13 is a vertical sectional view of a uranium 235 solution reactorcapable of operating continuously at substantial power for example 10kw.; and,

FIG. 14 is a vertical sectional view of the device of FIG. 13 taken in aplane at a right angle to the plane of the section in FIG. 13.

Referring first to FIGS. 1 and 2, a reactor tank 10 of spherical form isprovided approximately 12 inches in diameter and having a volume of14.95 liters, made of type 347 18-8 stainless steel, which issufficiently thin, for example ,4, inch thick, to absorb but minoramounts of neutrons. The sphere is made from two spun hemispheres with aV inch equatorial flare, and the hemisphere flares are welded together.Polar flares are also provided, to one of which is welded a top pipe 11.A bottom pipe 12 is welded to the other flare. The top pipe is 1 /2inches inside diameter with a inch wall and the bottom pipe is /1 inchoutside diameter with a inch wall. Unless otherwise specifiedhereinafter, all piping in the solution system is of stainless steel.

Referring first to bottom pipe 12, this pipe extends downwardly througha heavy frame base 14 and then through the top of an inverted conicalpan 15 to terminate inside thereof just above the bottom point of thepan. Pan 15 is supported on risers 16, which also partially support base14. Pan 15 can be emptied by a dump pipe 17 under the control of a dumpvalve 19 having an extension handle 20. A funnel 21 is provided throughwhich contents of sphere 10 and pan 15 can be conducted into a sump 22,when dump valve 19 is open. In view of the neutronic reactivity of thesolution to be used in reactor tank 10, tank 22 may be provided withneutron absorbers such as cadmium baffles 24, to prevent neutronicreaction therein.

Top pipe 11 extends upwardly through a cross-frame member 25, thiscross-frame member being supported by uprights 26 resting on frame base14. Above crossframe member 25 upper pipe 11 terminates in an expandedportion 27 provided with a removable cap 29. An overflow pipe 3 isprovided leading outwardly from expanded portion 27. The remainder ofthe liquid handling system will be explained later.

Inasmuch as a very considerable weight will be placed on base 14, base14 is additionally supported by base uprights 32. A reflector base 34formed from carefully machined graphite bricks is piled on base 14, thisgraphite being of high neutronic purity. Resting on graphite base 34 andsurrounding reactor tank is a reflector 35 of beryllium oxide brickshaving a density of about 2.7 gms./cm. carefully finished to fittogether with a minimum of air spaces, of maximum neutronic purity, andwith bricks adjacent the reactor tank lll being shaped to the contour ofthe tank. The beryllium oxide reflector is roughly of spherical shape toprovide a neutron reflecting layer around the reactor tank. Beforeassembling the reflector around the reactor tank, means for detectingleaks in the tank are provided in the form of small, preferably nyloninsulated, copper wires 36 wound around the tank 10. While only a singlecircuit is shown, separate circuits can be used for the top, equator andthe bottom of the reactor tank, if desired. If a leak from the tankoccurs, the solution will saturate the insulation on the wire andgroundit to the reactor tank 10, as will be later described.Thermocouples may also be inserted in various positions around thereactor tank, as indicated by thermocouple 37 positioned adjacent thetop of the reactor tank 10.

As the reflector 35 is being assembled, two vertical tangential slotsare built into space slightly away from tank 1 0 in the reflector, awide control rod slot 40 and a safety rod slot 41 close to tank 10. Bothof these slots may be provided with an aluminum lining or scabbardattached to the equator of tank 10. Operating in the control rod slot 40is a control rod 42. The control rod 42 proper is a strip of .032 inchcadmium 34 inches long, wrapped around a hollow brass tube 11 inch indiameter and 34 inches in length, and is moved in a vertical directionwith a total length of motion of 40.7 inches by a control rod motor 44,the position of the rod being indicated by Selsyn repeater 45. Thedetails of this control rod mechanism is shown more in detail in FIGS. 3to 6, inclusive, and will next be described.

As shown in FIG. 3, a screw shaft 46 is mounted vertically in a screwshaft bearing 47 mounted on top frame member 48 and extends upwardly toreceive a spur gear 49 pressed against a shoulder 50 keyed to shaft 46by clutch spring 51 retained by end nut 52. Clutch spring 51 forcesclutch plate 53 against spur gear 49 and spur gear 49 against shoulder50. Spur gear 49 is driven by the reversible DC. control rod motor 44through pinion 55. Thus screw shaft 46 is rotated by the motor 44through a friction clutch drive. The lower portion of shoulder 50 isprovided with pinion teeth 56 engaging a driven spur gear 57 attached tothe shaft of the Selsyn repeater 45.

Extending downwardly from bearing 47, is a rod casing 59 terminating ina casing block 60, which also supports an upwardly extending inner tube61. Immediately inside of inner tube 61 is a control rod sheath 62 whichextends all the way from bearing 47 to the full desired extent ofcontrol rod motion in control rod slot 40. Sheath 62 is sealed at thebottom by a welded cap 64.

The control rod proper, as above described, consists of a cadmium layer65, sandwiched between inner and outer brass tubes 66 and 67,respectively, these tubes being attached at their upper end to a nut 69sliding inside of rod sheath 62 and prevented from turning by aprojection 70 entering aligned slots 71 in tubes 61 and 62. Nut 69 isthreaded on threads 72 cut on the portion of shaft 46 below bearing 47.Thus rotation of shaft 46 will raise and lower the control rod within awatertight sheath. This watertight construction is not important when aberyllium oxide reflecor is utilized, but is useful in case a liquidreflector, such as deuterium oxide (heavy water), is used.

The safety rod 76 (FIGS. 1 and 2) consists of a cadmium sheet .032 inchthick, 2 /2 inches wide and 42 inches long, sandwiched for strengthbetween two similar pieces of brass. In its bottom position, its lowerend extends 8 inches below the center of the tank 10. Normally anelectromagnet 77 holds the safety rod out of the reflector by means of asafety rod disc 79 of magnetic material attached directly to the top ofthe rod. Any interruption of current in the magnet, brought about eithermanually or by means of any of the safety circuits, later to bedescribed, will release the rod to fall freely into the reflector bygravity. A tripping switch 80 is provided just above the top position ofthe magnet 77 so that if the magnet should be lifted too high, thesafety rod will be dropped. In addition, upper and lower positionindicator switches 81 and 82, respectively, are provided so that the inor out position of the safety rod can be made known to the operator ofthe reactor.

The safety rod is raised and lowered as desired by a safety rod motor84operating a drum 85 winding a cable 86' attached to the electromagnet77. The safety rod motor 84 is a'standard reversible motor, and isoperated through circuit R which is connected to a switch 148 located inthe control room." The switch 148' is a standard switch having threepositions, an off position, an electromagnet raising position, and anelectromagnet lowering position. Limit switches 89 and 90 are provided,operated by a stop on the drum 85, to limit the top and bottomrespectively, of the safety rod travel. Limit switch 80 is an additionalsafeguard in case limit switch 89 does not operate to stop motor 84. Asliding brake 91 is provided on the safety rod to soften the blow on thestructure when the rod is dropped. It may consist of two brass barsclamped to the rod with the friction adjusted by a spring. Normally,this brake is about 4 inches from the top of the rod. The rod,therefore, falls freely when released by magnet 77 until the brake 91hits cross-frame member 25 after which the rod has to slide betweenthebrass bars with some friction. A stop 92 is provided near the top of thestructure to prevent the brake from rising all the way, and for the last4 inches the safety rod has to slide between the brass bars, thusresetting the position of brake 91.

Certain other safeguards are attached to the system as so far described,and while their position will be indicated here, their functions will betaken up later. For example, immediately below the expanded portion 27on the top of the upper pipe 11 are a pair of solution contact switches95 and 96, switch 95 being slightly lower than switch 96. These switchesare used to monitor the upper level of solution in the reactor system.Such switches ground -a central electrode to the system through thesolution, which is electrically conductive. A lower level indicatorswitch 97 of similar type is provided on lower pipe 12 just above thetop of conical pan 15. The top of pan 15 is provided with a pan airsupply line 99 and an electrically operated air release valve 100, and alevel indicator 101. Pan 15 is also provided with a pan thermocouple 102for determining the temperature of the liquid in the pan 15.

Neutron monitoring ionization chambers are also provided. A pair of BFionization chambers 105 and 106 (FIG. 2) are provided outside of thereflector 35 and positioned behind a lead shield 107 where the chamberswill still receive a sufiicient neutron density during operation of thereactor to give proper monitoring of the neutron reactivity. Anadditional ionization chamber 109 (FIG. 1) is provided adjacent pan 15to monitor the radiation activity of the liquid in this pan.

The liquid handling system as shown in FIG. 7, will next be referred to.Inasmuch as one preferred solution to be used in the reactor is a uranylsulphate solution in ordinary water. with the uranium 235 content of theuranium much higher than in natural uranium, it is important thatevaporation from the solution be controlled so that the solution beinghandled may remain substantially constant in concentration during useand therefore is separated from the operating air. To attain thisresult, and to fill reactor tank 10, a source of compressed air isprovided, arriving through airpipe 110 under the control of system inletvalve 111 and balloon filling valve 112 (FIG. 7). Between the two valvesa supply pipe 114 leads to an air reservoir 115. Attached to pipe 114 isan air pressure gauge 116, and an electrically operated air dump valve117. In the interior of air reservoir 115 are positioned flexibleballoons 119 connected to pan air supply pipe 99 (FIG. I) line and to amanometer 121 having a liquid level indicator switch 122 therein. It ispreferred that valves 111 and 112, air pressure gauge 116, air dumpvalve 117 and manometer 121 be positioned within a control room, asindicated by enclosure line 125. All air lines are of OD. copper. Valvehandle 20 is also extended to this room.

It can be seen from the air line circuit so far described,

that air pressure can be applied to the top of a solution positioned inpan through the medium of the balloons 119, and that the system andballoons can be originally charged with air through valve 112.

At the top of the system, overflow pipe 30 (FIG. 7) leads to an overflowtank 126, in the bottom of which is positioned a liquid indicator switch127. Air from overflow tank 126 passes through airline 128 to outletballoon 129 prevented from collapse by internal umbrella frame 130, thussealing the system from the outside air. Thus by applying air pressureto the solution in pan 15, this solution can be forced upwardly throughpipe 12 into the reaction tank 10 and beyond tank 10 into upper pipe 11.

FIGURE 7 also illustrates the unreflected species of neutronic reactorclaimed below, all reflecting material having been removed to a distancesufiicient to render it ineffective. It is apparent that the size of thespherical tank 10 may be increased to a size appropriate to contain thenecessarily larger amount of fuel discussed below, specifically having aradius of 27.3 centimeters and containing 2920 grams of uranium 235. Itis also apparent that the other elements shown in FIGURE 7 and describedabove will require conventional structural support.

The operation of the reactor will be described after the electricalmonitor system has been described, as shown in FIG. 8, which is highlydiagrammatic and reduced to lowest terms. The right hand side of thediagram shows the instruments in the control room, the left hand sideindicates the circuit. The letters denote the continuity of the circuitswith those shown in other figures.

The three BF ionization chambers, i.e., chambers and 106, placedadjacent the reactor, see FIGURE 2, and the chamber 109 placed adjacentthe pan 15, see FIGURE 1, energize D.C. amplifiers 135, 136 and 139respectively. The outputs from these amplifiers operate respective monitoring galvanometers a, 136a, and 139a, to indicate radiation values.The output circuit from these amplifiers 135 and 136 also pass throughtripping circuit relays as indicated by numeral 140. The trippingimpulse, carried on trip line 141, operates a relay in a safety rodmagnet power supply indicated by numeral 142, to break current carriedin electromagnet circuit A, connected to electromagnet 77 holding up thesafety rod 76. The safety rod magnet power supply 142 can also beinterrupted by a hand switch 144, and by the upper safety rod limitswitch 80 through circuit B. The tripping impulse from the trippingcircuit also is passed through a time delay relay (about 2 secondsdelay) indicated by numeral 145, the impulse then travelling along line146 to operate a solenoid air release circuit indicated by numeral 147,the output of which is carried by circuit C, to the electricallyoperated air release valve 117, which is in the control room. Thesolenoid air release circuit 147 opening air release valve 117, is alsooperated by overflow contact switch 127 (FIG. 7) through circuit D; byleak detector wire 36 through circuit E (FIG. 1); and by upper liquidlevel switch 96, through circuit F (FIGS. 1 and 7). It will be noted inthis respect that the overflow contact switch 127 backs up upper levelswitch 96 as a safeguard in case switch 96 may not operate properly, as,for example, When the solution is blown over into overflow tank 126 byan air bubble, etc.

The control rod motor 44 is operated by in and out switch 150 throughcircuit G and Selsyn 45 drives a control rod depth indicator 151 in thecontrol room through circuit H.

The various thermocouples such as, for example, thermocouples 37 and 102operate temperature indicators 152 and 154 respectively, in the controlroom through circuits I and I respectively. Further, in order to ensureemergency release of the air pressure in the system, in

opening manual emergency switch 155. Normally the valve 100 is heldclosed by power fromthe mains through circuit K, but upon failure ofpower in the mains the valve is kept closed by a separate D.C. source156.

',All other safety circuits are arranged to operate upon .power failurein the mains. Safety rod position indicating switches 82 and 81operlevel switch 101 operates indicator 160 in the control room throughcircuit N. The liquid level indicator switch 122 in manometer 121 (FIG.7) operates lamp 161 through circuit 0, and the high and low solutionlevels in the system are indicated by lamp 162, operated by solutionswitch 95 through circuit P for the high position, and by lamp 164operated from solution switch 97 through circuit Q for the low position.Solution position indicating switches 95, 96, 97 and 101 are operatedWith circuits normally open to prevent electrolysis of the solution andare checked by push buttons.

Having thus described the reactor and the control and safety systems,the liquid handling system and the electrical monitoring and operatingsystem, the operation of the device will be described considering firstthat the reactive composition to be used in the reactor is a uranylsulphate solution having a suificient uranium 235 concentration to causethe system to be chain reacting when it has filled the 12 inch sphere.This uranyl sulphate solution is to be stored in pan 15 and level switch101 will detect solution level 124, see FIGURE 7, thereby indicating ifthere is suflicient solution in the pan to fill the reactor tank andpipes. To describe the process of filling the reactor tank andinitiating the reaction, reference is made to FIG. 7. The control andsafety rods are fully inserted. Then the balloons 119 are filled by useof the valve 112 which is then closed. Air pressure is then admitted toair reservoir 115 through valve 111, compressing the air in balloons119, causing the pressure to be transferred to the top of the solutionin the pan 15. The solution then is forced upwardly in pipe 12, intoreactor tank 10 and then into upper pipe '11. The progress of the riseof the solution through the system may be checked by watching manometer121 which is calibrated to give a rough position of the liquid level asit is rising. This manometer can also be used to detect air leaks in thesystem. The upper solution switch 95, mounted on upper pipe 11,indicates when the solution reaches that level, which is the normaloperating level of the solution in the system.

It will be noted that the solution will stay at this level only ifelectromagnetic air release valves 117 and 100 remain closed. 'If eitherof these valves are opened, the air pressure on the solution is releasedand this release will permit the solution to flow back into conical pan15, and out of the tank 10, by gravity. It will be noted from FIG. 8that the solution will be dumped by operation of valve 117 throughcircuit C, when anyone of a number of things happen. First, if thesolution level rises beyond upper solution level switch 96, second ifthere is an overflow into overflow tank 126, and third, if there is aleak in the tank itself, such as would saturate the nylon covering ofthe Wire 36 Wound around the reaction tank. Furthermore, valve 117 willalso be operated two seconds after the tripping circuit '140 has createdan impulse to drop the safety rod. The delay in this case is to preventthe solution from being dumped if the entrance of the safety rod intothe reflector properly stops a rise in neutron density, as would beindicated by either of the ionization chambers 105 or 106. If the safetyrod does not stop the rise in neutron'density after 2 seconds, thesolution is then automatically dumped out of'the sphere. 'Finally, ifnone of these automatic precautions operate, the manual emergency dumpcircuit K can be operated to release the air through the solenoidoperated air valve 100. This last operation is. only used'as a finalemergency procedure, as fresh air must then be supplied to the entiresystem. All dumping circuits except the'emer'gency circuit K arearranged so that if power supply fails, solenoid valve 117 will open.

As final precaution, if the solution is too highly radioactive asindicated by ionization chamber 109, and

galvanometer 139a,the conical pan'15 itself can be dumped by use of themanually operated valve handle 20, prcferablyextended to :the controlroom so that the solution can be conducted outside the operating roominto storage tank 22. The cadmium baflies 24, being strong neutronabsorbers, effectively prevent any possibility of the chain reactiontaking place in tank 22.

A discussion of some of the nuclear aspects of the system will be givenprior to describing the start-up of the operation of the device.

While the entire volume of the solution is normally stored in conicaltank 15, no chain reaction will take place therein for several reasons.First, the sphere is the most eflicient shape for a neutronic reactor,whereas the conical shape is not. Second, neutronic reactors of smallsize have an extremely high neutron leakage factor. When a reflector isused around tank 10, critical mass can be obtained with a lowerconcentration of uranium 235, because the reflector returns neutrons tothe solution and very elfectively reduces the amount of uranium 235required in the tank 10 to cause the chain reaction to be attained.Since conical pan 15 has no reflector, most of the escaping neutrons arelost and do not return. In consequence, no chain reaction takes place inpan 15.

However, the solution in pan 15 can become highly radioactive afteroperation of the device as a neutronic reactor, due to the accumulationof radioactive fission products therein. Ionization chamber 109 is usedto monitor this radioactivity and if it becomes too high, the solutionmay have to be drained into storage tank 22 until the radioactivitydecays to a safe handling value. Alternatively an auxiliary tank may beprovided as a substitute for tank 15.

Other features should be pointed out. It will be noted that neither thesafety rod nor the control rod enters the reactor tank 10. Smallreactors such as shown and described herein have such high neutronleakage that they usually are not of critical size without a reflectorand are dependent for proper operation for a given size, concentrationand shape, on the eflicient action of the reflector. In such a smallreactor the insertion of neutron absorbers even in the reflector outsideof the reactor tank will prevent the reflector from returning suflicientneutrons to keep the chain reaction sustained with the reactor having amass that would be critical if it were not for the absorption in thereflector. This affords a very simple and effective method of controlwithout insertion of neutron absorbers into the reacting portion of thereactor.

Of the uranium salts, UO SO is preferred for use in the reactor insteadof, for example, uranyl nitrate, first, because there is less unwantedneutron absorption with the sulphate than there is with the nitrate,and, second, the sulphate is more soluble in water than the nitrate.Furthermore, 18-8 stainless steel has showed extremely low corrosionrates after being in contact with UO SO solutions from 1 to 2 weeks.Consequently, all portions of the system which are to come into contactwith UO SO are pickled with normal 3 M UO SO solution for firom 1 to 3weeks before starting operations.

In starting up the device for the first time, a sufiicient amount ofdistilled water is placed in conical pan 15 to properly fill the reactortank and its attached pipes to the proper operating level as indicatedby solution switch 95. Uranyl sulphate containing isotope uranium 235 tothe point where the average composition of the material is about 14.7percent uranium 235, as indicated by mass spectrometer analysis, isadded to 1 or 2 liters of the water withdrawn from the conicalreservoir, and dissolved therein. The resulting solution is replaced inthe tank 15 and stirred. as, for example, with an electric mixer,through a cap to prevent evaporation of the water. The solution is thenrun up and down in the reactor tank 10' several times, While the controlrods are in, to improve the mixing. When the neutron counting'rate, asindicated by monitoring counters and 106 does not change with eachsuccessive filling of the tank 10, the solution is adequately. mixed,This method of adding the salt increases 17 the total volume of thesolution at each step', and to avoid accumulation of too much excesssolution, some of the solution is removed during the addition,evaporated, and the recovered UO SO made ready for further use.

To establish a chain reaction uranyl sulphate is added in the mannerdescribed, until critical conditions are reached, i.e., where theneutron reproduction ratio in the reactor tank equals unity. With the 12inch reactor tank, critical conditions are obtained with about 570 gramsof uranium 235.

With the cadmium control rod partly in, some uranium 235, such asapproximately 1.8 grams, is removed from the reactor tank so that thecontrol rod is for practical purposes in its full out position withcritical conditions prevailing. The control rod is then calibrated interms of mass of uranium 235 in the reactor tank by adding uranium 235(as UO SO and determining the new critical posit-ion of the control rod,i.e., the position where the neutron reproduction ratio is unity. FIG.11 shows a graph of the equivalent mass of uranium 235 in grams as thefunction of the control rod position in the reactor system. The controlrod in this case is all the way in at inch. The mass of uranium 235 inthe reactor tank can then be regulated to give critical conditions atsome intermediate position so that the control rod can be moved in orout of the reflector to provide shutdown of reactivity, or supercritioality where the neutron reproduction ratio will be over unity,respectively, as desired. FIGURE 12 shows the effect on the reciprocalperiod of adding uran-ium 235 after critical conditions have beenattained.

Because of the large coeflicient of expansion of water, several of thepertinent constants such as, for example, the age of the neutrons andthe thermal diffusion length (previously defined), depend on thetemperature. A very noticeable change in external leakage andconsequently, in critical position of the control rod, will take placeif the temperature changes even a few degrees. This tem perature effectin the presently described reactor is 7.3 grams equivalent of uranium235 per degree C.

For this reason, it is desirable to keep the reactor at a substantiallyconstant temperature while operating.

One manner by which the temperature may be stabilized is to enclose thereactor in a well insulated room and maintain the room at an elevatedtemperature, such as, for example, 39 C.

A satisfactory enclosure is shown in FIGS. 9 and 10 showing theapproximate size and location of the equipment therein. A room is formedflom heat insulating walls 170 to enclose a space 171 about 12 x 12' X12%. The enclosure is maintained at the elevated temperature of 39 C. byheaters 172 under the control of electronic circuits operated bythermocouples such as, for example, thermocouple junctions registeringthe difference in temperature between points in contact with the tank 10and a reference temperature as measured with a platinum resistancethermometer in an aluminum cylinder 174 attached to upper pipe 11 abovethe reflector. One junction of each thermocouple is also imbedded in thesame cylinder 174. Tests have shown that thermocouples positioned on theoutside of tank 10 will give a good measure of the solution inside.

Temperature gradients within the room are controlled by heater baflles175, and good mixing of the air is obtained by the use of aircirculating fans 176 provided with fan baflles 177. In this manner it ispossible to maintain the temperature of the solution and itssurroundings uniform to within a few hundredths of a degree 0., andconstant for many hours to within one hundredth of a degree C.

The reactor is then in condition to be operated at a low power level,such as, for example,'l watt. To obtain a desired power level aftercritical mass is obtained with the control rod partly in, the controlrod is retracted, so that the reactor is super critical. An exponentialrise in neutron density then occurs, at a rate determined by the 18amount of removal of the control rod. With the control rod just slightlyremoved from the critical position single doubling of the neutrondensity can be obtained in days or hours, if desired. Further removal ofthe rod will, of course, increase the rate of neutron density rise. Whenthe desired power level is reached, the rod is returned to positionwhere the reproduction ratio is again unity. With the temperaturestabilized, only minor movements of the control rod will be needed tomaintain the desired power level.

The reactor described is useful as a source of neutrons, and materialsto be irradiated can be placed in a reentrant tube extending downwardlyinto reactor tank 10 through pipe 11 from expanded portion 27 byremoving cap 29. Another of the main uses of the device described is forthe determination of the neutronic behavior of solutions containinglarge amounts of uranium 235 under various temperatures andconcentrations, while undergoing a self-sustaining chain reaction bynuclear fission.

FIGS. 13 and 14 inclusive show a similar type of reactor having as aprimary function the generation of both slow and fast neutrons by meansof a chain fission reaction.

Referring to these figures, the reactive solution may be stored instorage tank 200 positioned in tank chamber 201 surrounded by concretewalls 202 below ground. Storage tank 200 is larger than the reactor tankto be used, and thus the solution occupies only the lower portion of thetank, and is of unfavorable shape for the chain reaction. This fact,together with the fact that no reflector is used around tank 200,eifectively prevents the chain reaction from taking place therein.

The solution stored in tank 200 is pushed upwardly through solution pipe204 passing through a lead base 205 supported on base beams 206, into aspherical reactor tank 207, 12 inches in diameter, surrounded by acubical beryllium oxide reflector 208 three feet on a side, by airpressure in pipe 200 as in the first embodiment described. Horizontalinlet cooling water tubes 210 are extended around the reactor tank 207and then pass through the tank 207 to emerge as outlet water tube 211.

In the center of reactor tank 207 is a horizontal Scabbardtube 212 (FIG.14) into which a horizontal safety rod 214 enters. A vent pipe 215extends from the top of the reactor tank 207 outwardly parallel withinlet and outlet tubes 210 and 211, respectively. Tubes 210 and 211,safety rod 214, and vent pipe 215 all pass through a thick alternatelayer paraifin-bismuth shield 213 into a control room space 216. Thesolution can be dumped from reactor tank 207 by operation of valve 217,having an operating rod 217a extending into the control room space.

The opposite vertical race of reflector 208 is also provided with athick alternate layer parafiin-bismuth shield 213 (FIG. 13) pierced by afast neutron bore 218, entering reflector 208 adjacent reactor tank 207and opening into a space or room 219 wherein the fast neutrons may beutilized. The fast neutron bore maybe provided with bismuth filters 220,which pass fast neutrons but offer gamma ray shielding. Suitable means(not shown) may be provided to close off the bore if desired when thefast neutrons emanating therefrom are not in use.

t The other two vertical faces of reflector 208 (FIG. 14) are enclosed,first by inner vertical lead shields 221 and then by concrete walls 222.These walls 222 support column beams 224 extending across the top of thereflector at approximately ground level.

Erected on column beams 224 and centered over the top of reflector 208is a thermal neutron column 225 formed from graphite blocks of highneutronic purity. This thermal neutron column is preferably wider thanreflector 208 and is surrounded at its basal portion by concrete blocks226, and near the top portion by a lead shield 227. The upper face ofthe thermal neutron column is fiat, and upon this face may be placedmaterials or structures it is desired to expose to the action of thethermal neutrons. Thus the thermal neutrons can be utilized on the topof the thermal neutron column, and the fast neutrons utilized afteremerging from fast neutron bore 218.

Control is effected by a control rod 325 operating horizontally from thecontrol room 216 and entering the reflector 208 beneath the reactor tank207. Monitoring of the reaction is by ionization chamber 326 andappropriate indicator in the control room. Safety and check circuits maybe similar to those described for the embodiment previously described.

The operation of the embodiment just described is similar to that of thedevice first described, except that due to the coolant entering thereactor, it can operate continuously at a power output of 10 kw. with asubstantially uniform temperature. A power of 1 kw. can be obtained in atank 12 inches in diameter and containing about 670 grams of 235 in theform of UO SO This is about 100 grams more than is used in thepreviously described embodiment, and the extra uranium 235 is used inthe present embodiment to compensate for the neutron absorption in thecoolant tubes and safety rod sheath.

At the rated power of 10 kw. a flux of about 10 slow neutrons per cm.per second will be available on the surface of the thermal column 225,and a flux of about 10 fast neutrons per cm. per second will emerge fromthe fast neutron bore 218, both for use as desired.

In addition, solution from storage tank 200 can be withdrawn andreplaced through shielded pipe 228, and radioactive fission productsremoved, or uranium 235 added, etc., as desired. A powerful and compactneutron source has thus been provided.

While the embodiments above described have used uranium 235 as thefissionable isotope, it has been pointed outabove that plutonium 239 anduranium 233 can also housed.

It is apparent that the embodiment of FIGURES 13 nd 14 may be modifiedto be operated as an unreflected reactor in accordance with thedisclosure below of a. fuel solution of uranyl sulfate in an aqueoussolution containing 2.8 atoms of uranium 235 per molecule of water inwhich the uranium contains 12.5 percent uranium 235. As there indicated,the radius of such fuel in a.

spherical geometry is 27.3 centimeters, and 2920 grams of uranium 235are required. Such modifications would include elimination of theberyllia reflector 20 8, the paraflin-bismuth shield 213, lead shields205 and 227, concrete shields 22 6 and 202, thermal column 225 andcontrol rod 325. Safety rod 214 is of course modified to serve as thesole control rod. Suitable supporting structure is furnished for theremaining elements. It is here noted that the unre-flected embodiment ofFIG- URE 7 discussed above may be modified for control purposes bysubstituting for tank 10 thereof a tank similar to 207 of the FIGURE13-14 embodiment, complete with a suitable scabbard and control rod.

It has been found that a chain reaction can be established using asuspension or dispersion of at least about 10 grams of a fissionableisotope such as plutonium 239, uranium 233 or uranium 235 per cubiccentimeter of aqueous dispersion using ordinary water. The limitingminimum concentration is dependent to a substantial degree upon themoderator and even where the moderator has negligible neutron absorptionas is the case with D 0 the concentration must be at least 10- grams ofplutonium or other fissionable isotope per cc. of solution.

Above these minimums a substantial range of concentration is permissibleand this concentration may be as high as about 8-10 grams of uranium percc. of dispersion. However, it is rare that concentrations muchabove-about 2 grams of tissionable isotope per cc. of solution are used.Preferably the dispersion should he a true, solution which issubstantially free from undissolved suspended fissionable solids sincecontrol of the reaction is much easier in such a case.

In order to control the reaction without excessive effort it generallyis, preferred to maintain the solution or suspension at a substantiallyconstant concentration. While some variation is permissible, widevariation in concentration while the reaction proceeds makes control ofthe, chain reaction diflicult.

The amount of fissionable isotope which should be present in order. toestablish a self-sustaining neutron chain reaction depends to asubstantial degree upon the concentration ofthe fissionable isotope inthe moderator andv also upon. the neutron absorption characteristic ofthemoderator used. In general it can be said that the amount of uranium235 present should be at least about 300 grams with optimumconcentration and using either pure uranium 23 5. or uraniumconcentrates containing 5 to 10 percent or more ofuranium 235. The exactamount requiredwill also depend upon the fissionable isotope. whichisused. For example, it has been found that when plutonium 239 is usedas the fissionable isotope only about two-thirds of the weight ofisotope required foruranium- 235 is necessary.

The following tables indicate generally the trend. Table I tabulatesthequantities and critical size required" for a spherical reactorcontaining uranyl sulphate dis solved in water, when the reactor isprovided with an infinite D 0 reflector and when the uranium isenrichedto contain 12.5 percent by weight of uranium 235 the balancebeing uranium 238. In the table z denotes the number of atoms of=uranium235 present per moleculemixture. r denotes the radius of the sphere incenti meters andG denotes-thecritical quantity of 235 required in grams.

Table I Z 0 d, grnJem. r, em G, gm.

2. 0 1. 01s 32, 770 3.0 1.026 26.8 .524 3.9- 1. 034 20.1 848 4.1 1. 04017. 0 112 5. 4 1. 048 16.3 622 6.1 1. 054 15. a 582 6..8 1. 062 14. 6564 7. 5 1. 00s 14. 0 54s 9. 3 1.086 13. 0 537 10. 9 1. 102 12. 4 54716. 9 1.170 11.0 587 From the above table it will be apparent that asthe concentration of uranium 235 in a solution increases from 1X10 atomsof the uranium 235 per molecule of H 0 the critical radius of thereactor decreases and the critical mass of 235 decreases to a minimumsomewhat over 500 grams and thereafter the critical mass increases withincreasing concentration.

Where a neutron moderator which has less tendency than water to absorbneutrons is used, the critical mass for a chain reaction may besubstantially smaller. The following table tabulates the critical masswhich is required for various concentrations of a uranyl-plutonylsulphate solution in D 0 using an infinite D 0 reflector. Theconcentration of plutonium was 12.5 percent based upon the total, Weightofuraniumand plutonium. In the table Z and G are as defined above. V isthe critical volume inliters.

Frornthegabove Table II it is shown that a critical mass asv low asabout 200 grams of plutonium 239 is capable, of sustaining a reaction ina D 0 moderator.

1. A BARE NEUTRONIC REACTOR HAVING A LIQUID FUEL CONSISTING OF ASPHERICAL VESSEL HAVING A RADIUS OF AT LEAST ABOUT 6 INCHES, SAID RADIUSBEING DEFINED BY THE EQUATION: